• KOREAN JOURNAL OF PACKAGING SCIENCE & TECHNOLOGY
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• v.29 no.3
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• pp.203-209
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• 2023

The increasing saturation challenges in storage facilities for Low- and Intermediate-Level Radioactive Waste call for a more efficient storage approach. Consequently, we have developed a square-structured container that features a storage capacity approximately 20% greater than that of conventional drum-type containers. Considering the need to contain various radioactive wastes from nuclear power usage securely until they no longer pose a threat to human health or the environment, this study focuses on evaluating the sealing efficacy of the newly designed rectangular container using finite element analysis. Since radioactive waste containers typically do not experience external forces except under special circumstances, our analysis simulated the impact of an external force, assuming a fall scenario. After fastening the bolts, we examined the vertical stress distribution on the container by applying the calculated external force. The analysis confirms the container's stable seal.

• Economic and Environmental Geology
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• v.39 no.2 s.177
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• pp.173-180
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• 2006

The synthesis of polyphase waste form, which is an immobilization matrix fur the high level radioactive wastes, was performed with the mixed composition of garnet and spinel $(Gd_3Fe_5O_{12}+(Ni_xMn_{1-x})(Fe_yCr_{1-y})_2O_4)$ in the range of 1200 to $1400^{\circ}C$. The phases synthesized from all stoichiometric compositions were garnet, perovskite, and spinel. Especially, garnet was synthesized only in the composition of the highest content of Fe(y=0.9), whereas it was not synthesized in other compositions. This result indicated that the content of Fe was closely related to the formation of garnet. The composition of garnet revealed that the content of Gd was exceeded and that of Fe was depleted. Preferential distribution of elements in the phases can be attributed to the nonstoichiometric composition of garnet.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.3
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• pp.253-266
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• 2016

In accordance with the classification system of radioactive waste in Korea, all the disused sealed radioactive sources (DSRSs) fall under the category of EW, VLLW or LILW, and should be managed in compliance with the restrictions for the disposal method. In this study, the management and disposal method are drawn in consideration of half-life of radionuclides contained in the source and A/D value (i.e. the activity A of the source dividing by the D value for the relevant radionuclide, which is used to provide an initial ranking of relative risk for sources) in addition to the domestic classification scheme and disposal method, based on the characteristic analysis and review results of the management practices in IAEA and foreign countries. For all the DSRSs that are being stored (as of March 2015) in the centralized temporary disposal facility for radioisotope wastes, applicability of the derivation result is confirmed through performing the characteristic analysis and case studies for assessing quantity and volume of DSRSs to be managed by each method. However, the methodology derived from this study is not applicable to the following sources; i) DSRSs without information on the radioactivity, ii) DSRSs that are not possible to calculate the specific activity and/or the source-specific A/D value. Accordingly, it is essential to identify the inherent characteristics for each of DSRSs prior to implementation of this management and disposal method.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.3
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• pp.141-148
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• 2011

It is clarified in the radioactive waste transfer regulation that the concentration of radioactive waste for the major radio nuclide has to be examined when radioactive waste is guided to the radioactive waste stores. In case of the low level radioactive waste sample, the analytical results of radioactive waste concentration frequently show a value lower than minimum detectable activity (MDA). Since the MDA value basically depends on the amount of a sample, background value, measurement time, counting efficiency, and etc, it would be necessary to increase a sample amount with a intention of minimizing MDA. In order to measure a concentration of $^{129}I$ in low and medium level radioactive waste, $^{129}I$ was collected by using a distillation technique after leaching the simulated radioactive waste sample with a non-volatile acid. The recovery of $^{129}I$ measured was compared with that measured with column elution technique which is a conventional method using an anion-exchange resin. The recovery of inactive iodide by using the distillation method and column elution were found as $86.5{\pm}0.9%$ and $87.3{\pm}2.7%$, respectively. The recovery and MDA value calculated for distillation technique when 100 g of extracted solution of $^{129}I$ was taken, were found to be $84.6{\pm}1.6%$ and $1.2{\times}10^{-4}Bq/g$, respectively. Consequently, the proposed technique with simplified process lowered the MDA value more than 10 times compared to the column elution technique that has a disadvantage of limited sampling amount.

• The Korean Journal of Nuclear Medicine Technology
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• v.16 no.2
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• pp.12-17
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• 2012

Purpose : The patient's clothes and sheet after radioiodine therapy must be disposed of by related regulation. That must be disposed of as radioactive wastes, but that is reusing after radioactivity decay by keeping for the certain period of time. In general, The minimum storage period calculate by standard of take radioactive substance out of radiation controlled area based on measured surface contamination level. But the measurements of surface contamination level are able to differ by measurement method. In this paper, I wish to calculate the minimum storage period of patient's clothes and sheet after radioiodine therapy by measure nuclide concentration offered by the regulation on self-disposal of radioactive wastes. Materials and Methods : The whole area of patient's clothes and sheet measured 31 patients(male:9 patients, female:22 patients), who had radioiodine therapy(3.7 GBq:13 patients, 5.55 GBq:16 patients, 7.4 GBq:2 patients) from july 2011 to march 2012. The minimum storage period is calculated by the regulation on self-disposal of radioactive waste(100 Bq/g) and standard of take radioactive substance out of radiation controlled area(4 kBq/m2) Results : The minimum storage period of pillow sheet, upper uniform, lower uniform by standard of take radioactive substance out of radiation controlled area were each 4.6 days, 63days, 78 days. The minimum storage period of pillow sheet, upper uniform, lower uniform by the regulation on self-disposal of radioactive waste were each 18.1 days, 43 days, 62 days. Conclusion : We can verify that patient's clothes and sheet after radioiodine therapy exists a great deal of radioactive contamination. The minimum storage period calculation of patient's clothes and sheet is better suited to applying nuclide concentration offered by the regulation on self-disposal of radioactive waste. I recommend, To keep for at least 2 months of the patient's clothes and sheet contaminated radioactivity, for prevent contamination and unnecessary radiation exposure.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.2
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• pp.187-195
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• 2006

The regulation for the low and intermediate level radioactive waste to be transferred to the disposal facility, recently revised, require that radioactive waste generators should set up waste certification program to verify the radioactive waste conform to the waste acceptance criteria(WAC) before disposal. The radioactive waste disposal facility, scheduled to be constructed in Korea, will institute WAC for the wastes to be transferred to the facility. This WAC is expected to compose of the requirements for the radiological characterization, physical and chemical characterization, physical/chemical restriction, prohibited item, packaging, identification, labeling, and documentation. For the compliance with this regulation, The radioactive waste generators should verify that the waste meet WAC through performance of the waste certification program and are responsible for handing in all the certification documents to the disposal facility. This waste certification program plan was set up as a preliminary program for the certification of radioactive waste generated in Korea Atomic Energy Research Institute (KAERI) and should be further revised until preparation of WAC by disposal agent.

• Carbon letters
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• v.3 no.1
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• pp.6-12
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• 2002

A study on the electrosorption of $Co^{2+}$ and $Sr^{2+}$ ions onto a porous activated carbon fiber (ACF) was performed to treat radioactive liquid wastes resulting from chemical or electrochemical decontamination and to regenerate the spent carbon electrode. The result of batch electrosorption experiments showed that applied negative potential increased adsorption kinetics and capacity in comparison with open-circuit potential (OCP) adsorption for $Co^{2+}$ and $Sr^{2+}$ ions. The adsorbed $Co^{2+}$ and $Sr^{2+}$ ions are released from the carbon fiber by applying a positive potential on the electrode, showing the reversibility of the sorption process. The possibility of application of the electrosorption technique to the separation of radionuclides was examined. The result of a selective removal experiments of a single component from a mixed solution showed that perfect separation of $Co^{2+}$ and $Sr^{2+}$ ions was possible by the electrosorption process.

• Proceedings of the Korean Environmental Sciences Society Conference
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• 2005.05a
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• pp.267-271
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• 2005

This experiment was carried out at 450"C, which is relatively lower than the temperature for supercritical water oxidation (600-650$^{\circ}C$). In this experiment, the decomposition rates of various incombustible organic substances were very high. In addition, it was confirmed that hetero atoms existed in organic compounds and chlorine was neutralized by sodium(salt formation).However, to raise the decomposition rate, relatively large amount of sodium nitrate(3-4 times the equivalent weight) was required. When complete oxidation is intended as in the case with PCB, the amount of oxidizer and decomposition cost is important. But when vaporization reduction is required as in the case with nuclear wastes, the amount of radioactive wastes increases instead. But as can be seen in the result of XRD measurement, unreacted sodium nitrate remained unchanged. If oxidation reaction of organic substance simply depends on collision frequency, unreacted sodium nitrate can be recovered and reused, then oxidation equivalent weight would be sufficient. In the gas generated, toxic gas was not found. As the supercritical water medium has high reactivity, it is difficult to generate relatively low energy level SO$_{X}$, and NO$_{X}$.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.3
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• pp.213-229
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• 2005

Nuclides in radioactive wastes are assumed to be transported in the geosphere by groundwater and probably discharged into the biosphere. Quantitative evaluation of doses to human beings due to nuclide transport in the geosphere and through the various pathways in the biosphere is the final step of safety assessment of the radioactive waste repository. To calculate the flux to dose conversion factors (DCFs) for nuclides appearing at GBIs with their decay chains, a template ACBIO which is an AMBER case file based on mathematical model for the mass transfer coefficients between the compartments has been developed considering material balance among the compartments in biosphere and then implementing to AMBER, a general and flexible software tool that allows to build dynamic compartment models. An illustrative calculation with ACBIO is shown.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.90-95
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• 2003

Chemical management of annulus gas system is carried out for the purpose of ensuring the safety and reliability of the system via securing the integrity of the system, detecting the D$_2$O in-leakage of coolant and/or moderator, and reducing the radiation dose. Since the quality of CO_2$ gas, which is used as a filling gas for annulus gas system at CANDU plants, has a propound effect on the integrity of the system material and the radiation dose, CO_2$ gas of high quality is needed. If the quality of CO_2$ gas does not meet the specification, it may give rise to undesirable effect not only on the annulus gas system, but also on the environment due to the production of radioactive nuclei. Therefore, it is very important to check the impurities of CO_2$ gas. Based on this background, the inventories of C-14 and Ar-41 in CO_2$ gas that is supplied as annulus gas were estimated using the data on concentrations of the impurities of $CO_2$ such as C, N_2$ and Ar. The results of this study is expect to give useful information on optimization of CO_2$ impurities maintenance and management of gaseous radioactive wastes produced at CANDU plants.